• 한국방사선학회논문지
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• 제12권6호
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• pp.713-718
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• 2018

많은 선행 연구에서는 무연 차폐재를 제작하기 위하여 몬테카를로 시뮬레이션을 통해 방사선 차폐 능력과 경량화에 대한 가능성을 제시하고 있다. 하지만, 이는 바인더 및 미세 기공에 대한 구현이 어렵기에 제품화 공정에 필요한 정보를 충분히 제공하지 못하는 실정이다. 이에 본 연구에서는 제품화 공정에 요구되는 겔 페이스트에 대한 정보를 사전에 제공하기 위하여 스크린 프린팅 공법을 활용하여 충전율에 따른 방사선 차폐 능력에 대한 결과를 제시하였다. 본 연구에서는 방사선 차폐 능력을 평가하기 위해 IEC 61331-1: 2014와 KS A 4025에 부합하도록 실험 환경을 설계하였으며, 방사선 조사 조건은 KS A 4021 규격을 준용하여 총 여과 2.0 mmAl로 여과된 100 kVp를 이용하였다. 본 연구 결과, TVL를 기준으로 Pb $1270{\mu}m$, $BaSO_4$ $3035{\mu}m$, $Bi_2O_3$ $1849{\mu}m$, $WO_3$ $2631{\mu}m$에서 근사한 값으로 분석되었다. 또한, 충전율은 $BaSO_4$ 38.6%, $Bi_2O_3$ 27.1%, $WO_3$ 30.15%로 분석되었다. 하지만, 차후 저온고압 성형을 적용한다면 충전율을 높이면서도 기공률을 낮춤으로서 방사선 차폐 능력의 개선이 충분히 가능할 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3470-3477
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• 2022

Soils are porous materials with high shielding capability to attenuate gamma and X-rays. The disposal of radionuclides throughout the soil profile can expose the living organisms to ionizing radiation. Thus, studies aiming to analyze the shielding properties of the soils are of particular interest for radiation shielding. Investigations on evaluating the shielding capabilities of highly weathered soils are still scarce, meaning that additional research is necessary to check their efficiency to attenuate radiation. In this study, the radiation shielding properties of contrasting soils were evaluated. The radiation interaction parameters assessed were attenuation coefficients, mean free path, and half- and tenth-value layers. At low photon energies, the photoelectric absorption contribution to the attenuation coefficient predominated, while at intermediate and high photon energies, the incoherent scattering and pair production were the dominant effects. Soils with the highest densities presented the best shielding properties, regardless of their chemical compositions. Increases in the attenuation coefficient and decreases in shielding parameters of the soils were associated with increases in clay, Fe₂O₃, Al₂O₃, and TiO₂ amounts. In addition, this paper provides a comprehensive description of the shielding properties of weathered soils showing the importance of their granulometric fractions and oxides to the attenuation of the radiation.

• 한국임상수의학회지
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• 제41권3호
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• pp.157-164
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• 2024

The standard radiation protection method in the angiography suite involves the use of a thyroid shield, a lead apron, and lead glasses. However, exposure to substantial amounts of ionizing radiation can cause cataracts, tumors, and skin erythema. A newly developed curtain-type radiation protection device consists of a curtain drape composed of a five-layer bismuth and lead acrylic head-shielding plate, with both bearing an equivalent 0.25 mm lead thickness. In this study, a quality assurance phantom was used as the patient to create radiation scatter from the radiographic source, and an anthropomorphic mannequin phantom was used as the interventionalist to measure the radiation dose at seven different anatomical locations. Thermoluminescent dosimeters were used to measure the radiation dose. The experimental groups consisted of all-sided or one-sided curtain set-ups, the presence or absence of a conventional shielding system, and the orientation of beam irradiation. Consequently, the curtain-type radiation protection device exhibited better radiation protection range and capabilities than conventional radiation protection systems, especially in safeguarding the forehead, eyes, arms, and feet, with minimal radiation exposure. Moreover, the mean shielding ratios of the conventional shielding system and curtain-type radiation protection device were measured at 51.94% and 93.86%, respectively. Additionally, no significant decrease in the radiation protection range or capability was observed, even with changes in the beam orientation or one-sided protection. Compared with a conventional shielding system, the curtain-type radiation protection device decreased radiation exposure doses and improved comfort. Therefore, it is a potential new radiation protection device for veterinary interventional procedures.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3058-3067
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• 2021

Sodium zinc borate glasses doped with dysprosium and modified with different concentrations of barium oxide (0-50 mol %) were fabricated using the melting quenching technique. The structural properties of the prepared glass systems were characterized using XRD and FTIR methods. The absorption spectra of the prepared glasses were measured to determine their energy gap and their related optical properties. The density of the glasses and other physical parameters were also reported. Additionally, with the help of Photon Shielding and Dosimetry (PSD) software, we investigated the radiation shielding parameters of the prepared glass systems at different energy values. It was found that an increase in the density of the glasses by increasing the concentration of BaO significantly improved the gamma ray shielding ability of the samples. For practical results, a compatible irradiation set up was designed to check the shielding capability of the obtained glasses using a gamma ray source at 662 keV. The experimentally obtained results strongly agreed with the data obtained by PDS software at the same energy. These results demonstrated that the investigated glass system is a good candidate for several radiation shielding applications when comparing it with other commercial shielding glasses and concretes.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3429-3439
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• 2022

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this paper, the radiation shielding capability of the MCS code is validated by simulating some selected SINBAD (Shielding Integral Benchmark Archive and Database) experiments. The whole validation was performed in two ways. Firstly, the functionality and computational rationality of the MCS code was verified by comparing the simulation results with those of MCNP code. Secondly, the validity and computational accuracy of the MCS code was confirmed by comparing the simulation results with the experimental results of SINBAD. The simulation results of the MCS code are highly consistent with the those of the MCNP code, and they are within the 2σ error bound of the experiment results. It shows that the calculation results of the MCS code are reliable when simulating the radiation shielding problems.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.957-962
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• 2018

The gamma radiation shielding capability (GRSC) of two clay-materials (Ball clay and Kaolin)of Southwestern Nigeria ($7.49^{\circ}N$, $4.55^{\circ}E$) have been investigated by determine theoretically and experimentally the mass attenuation coefficient, ${\mu}/{\rho}(cm^2g^{-1})$ of the clay materials at photon energies of 609.31, 1120.29, 1173.20, 1238.11, 1332.50 and 1764.49 keV emitted from $^{214}Bi$ ore and $^{60}Co$ point source. The mass attenuation coefficients were theoretically evaluated using the elemental compositions of the clay-materials obtained by Particle-Induced X-ray Emission (PIXE) elemental analysis technique as input data for WinXCom software. While gamma ray transmission experiment using Hyper Pure Germanium (HPGe) spectrometer detector to experimentally determine the mass attenuation coefficients, ${\mu}/{\rho}(cm^2g^{-1})$ of the samples. The experimental results are in good agreement with the theoretical calculations of WinXCom software. Linear attenuation coefficient (${\mu}$), half value layer (HVL) and mean free path (MFP) were also evaluated using the obtained ${\mu}/{\rho}$ values for the investigated samples. The GRSC of the selected clay-materials have been compared with other studied shielding materials. The cognizance of various factors such as availability, thermo-chemical stability and water retaining ability by the clay-samples can be analyzed for efficacy of the material for their GRSC.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• 한국방사선학회논문지
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• 제14권6호
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• pp.733-740
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• 2020

의료종사자의 피폭선량을 줄이기 위한 최근 연구에서, 방사선이 산란될 때 발생하는 광전효과를 이용하여 방사선치료실 입구에서의 선량을 줄이는 방법이 제안되었다. 이 방법은 특히 저에너지 광자에 효과적이기 때문에 본 연구에서는 몬테카를로 시뮬레이션을 이용하여 슬릿형태 구조물의 일반촬영실 산란 엑스선에 대한 차폐성능을 평가하였다. 두께 2 mm, 폭 50 mm, 길이 900 mm인 판을 2 mm 간격으로 수평 적재하는 형태의 슬릿형태 구조물은 알루미늄에 비해 철 또는 납으로 만드는 경우 차폐효과가 뛰어났다. 재질을 철로 한정한 경우 선원과 관심구역 사이에서 결정된 구조물의 설치위치는 차폐효과와 무관했으며, 판의 폭은 차폐효과에 비례했다. 폭 50 mm 철판을 사용한 경우 산란선이 직접 발생하는 바닥 및 환자의 높이를 제외하면 약 99.9% 또는 그 이상의 차폐효과가 있었다.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1557-1564
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• 2020

In this paper, 3D model-based interactive gamma ray shielding package (MIGSHIELD) is developed in virtual reality platform for windows operating system. In MIGSHIELD, the computational methodology is based on point kernel algorithm (PK), several key parameters of PK are obtained using new technique and new methods. MIGSHIELD has interactive capability with virtual world. The main features made in the MIGSHIELD are (i) handling of physical information from virtual world, (ii) handling of arbitrary shapes radioactive source, (iii) calculating the mean free path of gamma ray, (iv) providing interactive function between PK and virtual world, (v) making better use of PK for virtual simulation, (vi) plug and play. The developed package will be of immense use for calculations involving radiation dose assessment in nuclear safety and contributing to fast radiation simulation for virtual nuclear facilities.